Model Comparison of Passive Compact-Molten Salt Reactor Neutronic Design Using MCNP6 and Serpent-2

Abstract

Passive Compact Molten Salt Reactor (PCMSR) is a thermal breeder molten salt reactor (MSR) developed in Universitas Gadjah Mada, Indonesia, run in thorium fuel cycle. Its design was initially developed using deterministic code SRAC2006 but has never been compared with other codes. This paper attempts to compare PCMSR neutronic design using Monte Carlo codes MCNP6 and Serpent-2 with ENDF B/VII.0 continuous neutron cross-section library. The reactor was run in a pure thorium fuel cycle with lithium fluoride as its carrier salt. The analyzed parameters were effective multiplication factor (k eff ), temperature coefficient of reactivity (TCR), void coefficient of reactivity (VCR), and conversion ratio (CR). The result shows that there are several important discrepancies between the original calculation and this research. The Monte Carlo calculations implied that PCMSR core was able to be critical using lower fissile concentration than previously designed, but failed to reach CR above unity. While the TCR value was found to be negative, the VCR value was positive up until the 10 % void fraction. The PCMSR core suffered from ineffective neutron moderation and high neutron leakage. These findings imply that the previous PCMSR neutronic design is inaccurate. For PCMSR to be able to operate as a thermal breeder MSR, geometrical modifications must be performed to improve neutron moderation and reduce neutron leakage.

Neutronic Design Modification of Passive Compact-Molten Salt Reactor

“…Although MCNP may not be the most suitable simulation tool for MSR due to its decoupling from thermal-hydraulic calculations, it has nonetheless been used for simulating various MSR designs, such as MSBR [6,31,32], MSR-FUJI [9], TMSR-500 [33][34][35], and Integral Molten Salt Reactor (IMSR) [36], with good agreement to the reference. The original PCMSR design was also simulated with MCNP and found to be in good agreement with the Serpent-2 code [23].…”

“…However, most design studies of PCMSR were conducted using one-fluid core due to its design simplicity [19][20][21][22]. The initial design calculation was also one-fluid [19], in which neutronic economy was proven to be deficient by our recent study [23]. PCMSR core is moderated by graphite which can be replaced without shutting down the reactor.…”

“…Due to sparing development, however, PCMSR has no fixed core design [19][20][21][22]. Our previous study even concluded that the initial PCMSR core design was incapable of attaining thermal breeding, i.e., having neutronic deficiency [23].…”

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

“…MCNP is capable of modelling a complex geometry without simplification, allowing more accurate prediction of the reactor physics characteristics. Previously, MCNP was used to calculate neutronic parameters of MSBR [36,37], Integral Molten Salt Reactor (IMSR) [38], MSR-FUJI [39], TMSR-500 [40,41], and PCMSR [42]. Therefore, MCNP can be considered to be suitable to calculate the neutronic parameters of an MSR, despite the calculation was performed in a quasistatic condition.…”